[Federal Register Volume 68, Number 198 (Tuesday, October 14, 2003)]
[Notices]
[Pages 59212-59229]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-25742]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of 
1954, as amended (the Act), to require the Commission to publish notice 
of any amendments issued, or proposed to be issued, under a new 
provision of section 189 of the Act. This provision grants the 
Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, September 19, 2003, through October 2, 
2003. The last biweekly notice was published on September 30, 2003 (68 
FR 56340).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission

[[Page 59213]]

take this action, it will publish in the Federal Register a notice of 
issuance and provide for opportunity for a hearing after issuance. The 
Commission expects that the need to take this action will occur very 
infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By November 13, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North,

[[Page 59214]]

Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to [email protected].

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: July 14, 2003.
    Description of amendment request: The proposed change involves the 
extension from 1 hour to 24 hours of the completion time (CT) for 
Condition B of Technical Specification (TS) 3.5.1, which defines 
requirements for accumulators. Accumulators are part of the emergency 
core cooling system and consist of tanks partially filled with borated 
water and pressurized with nitrogen gas. The contents of the tank are 
discharged to the reactor coolant system if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the 
accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore 
an accumulator to operable status when it has been declared inoperable 
for a reason other than the boron concentration of the water in the 
accumulator not being within the required range. This change was 
proposed by the Westinghouse Owners Group participants in the Technical 
Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is 
supported by NRC-approved topical report WCAP-15049-A, ``Risk-Informed 
Evaluation of an Extension to Accumulator Completion Times,'' submitted 
on May 18, 1999. The NRC staff issued a notice of opportunity for 
comment in the Federal Register on July 15, 2002 (67 FR 46542), on 
possible amendments concerning TSTF-370, including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on March 12, 2003 (68 FR 11880). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated July 14, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The basis for the accumulator limiting condition for operation 
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a 
sufficient volume of borated water will be immediately forced into 
the core through each of the cold legs in the event the RCS pressure 
falls below the pressure of the accumulators, thereby providing the 
initial cooling mechanism during large RCS pipe ruptures. As 
described in Section 9.2 of WCAP-15049-A, the proposed change will 
allow plant operation with an inoperable accumulator for up to 24 
hours, instead of 1 hour, before the plant would be required to 
begin shutting down. The impact of the increase in the accumulator 
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total 
plant core damage frequency (CDF) less than 1.0E-03/yr. The 
incremental conditional core damage probabilities calculated in 
WCAP-15049-A for the accumulator CT increase meet the criterion of 
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049-A, design 
basis accumulator success criteria are not considered necessary to 
mitigate large break loss-of-coolant accident (LOCA) events, and 
were only included in the WCAP-15049-A evaluation as a worst case 
data point. In addition, WCAP-15049-A states that the NRC has 
indicated that an incremental conditional core damage frequency 
(ICCDP) greater than 5E-07 does not necessarily mean the change is 
unacceptable.
    The proposed technical specification change does not involve any 
hardware changes nor does it affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, engineered safety feature (ESF) actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this 
proposed technical specification CT increase. All safety systems 
still function in the same manner and there is no additional 
reliance on additional systems or procedures. The proposed 
accumulator CT increase has a very small impact on core damage 
frequency. The WCAP-15049-A evaluation demonstrates that the small 
increase in risk due to increasing the CT for an inoperable 
accumulator is within the acceptance criteria provided in RGs 1.174 
and 1.177. No new accidents or transients can be introduced with the 
requested change and the likelihood of an accident or transient is 
not impacted.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be caused as a result 
of the proposed technical specification change. No new failure mode 
has been created and no new equipment performance burdens are 
imposed.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits.
    The basis for the accumulator LCO, as discussed in Bases Section 
3.5.1, is to ensure that a sufficient volume of borated water will 
be immediately forced into the core through each of the cold legs in 
the event the RCS pressure falls below the pressure of the 
accumulators, thereby providing the initial cooling mechanism during 
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable 
accumulator for up to 24 hours, instead of 1 hour, before the plant 
would be required to begin shutting down. The impact of this on 
plant risk was evaluated and found to be very small. That is, 
increasing the time the accumulators will be unavailable to respond 
to a large LOCA event, assuming accumulators are needed to mitigate 
the design basis event, has a very small impact on plant risk. Since 
the frequency of a design basis large LOCA (a large LOCA with loss 
of offsite power) would be significantly lower than the large LOCA 
frequency of the WCAP-15049-A evaluation, the impact of increasing 
the accumulator CT from 1 hour to 24 hours on plant risk due to a 
design basis large LOCA would be significantly less than the plant 
risk increase presented in the WCAP-15049-A evaluation.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.


[[Page 59215]]


    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: October 16, 2001; as supplemented by 
letters dated May 20, September 12, and November 21, 2002; and January 
27, and September 22, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to incorporate changes 
resulting from the use of an alternate source term (AST) and the 
implementation of several plant modifications. Publication of the 
Proposed No Significant Hazards Consideration Determination and 
Opportunity for Hearing for the October 16, 2001, submittal appeared in 
the Federal Register on January 22, 2002, (67 FR 2922). The September 
22, 2003, submittal contained a revised No Significant Hazards 
Consideration Determination. The September 22, 2003, submittal includes 
(1) Implementing the AST for accident analysis as described in 
Regulatory Guide 1.183; (2) relaxing the TS for the penetration room 
ventilation system (PRVS) and the spent fuel pool ventilation system 
(SFPVS) because these systems are no longer credited for control room 
and offsite doses; (3) revising the control room ventilation system 
(CRVS) to allow for a one-time completion extension to support 
implementation of the control room intake/booster fan modification; (4) 
lowering the reactor building leakage rate from 0.25 weight percent per 
day to 0.20 weight percent per day; (5) revising the ventilation filter 
testing program radioactive methyl iodide removal acceptance criterion 
for PRVS, SFPVS, and CRVS booster fan trains; and (6) adoption of TS 
Task Force (TSTF)-51.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability of consequences of an accident 
previously evaluated.
    The AST [alternate source term] and those plant systems affected 
by implementing the proposed changes to the TS [technical 
specifications] are not assumed to initiate design basis accidents. 
The AST does not affect the design or operations of the facility. 
Rather, the AST is used to evaluate the consequences of a postulated 
accident. The implementation of the AST has been evaluated in the 
revisions to the analysis of the design basis accident for ONS 
[Oconee Nuclear Station]. Based on the results of these analyses, it 
has been demonstrated that, with the requested changes, the dose 
consequences of these events meet the acceptance criteria of 10 CFR 
50.67 and RG [Regulatory Guide] 1.183. Therefore, the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The AST and those plant systems affected by implementing the 
proposed changes to the TS are not assumed to initiate design basis 
accidents. The systems affected by the changes are used to mitigate 
the consequences of an accident that has already occurred. The 
proposed TS changes and modifications do not significantly affect 
the mitigative function of these systems. Consequently, these 
systems do not alter the nature of events postulated in the Safety 
Analysis Report nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The implementation of the AST, proposed changes to the TS and 
implementation of the proposed modifications have been evaluated in 
the revisions to the analysis of the consequences of the design 
basis accidents for the ONS. Based on the results of these analyses, 
it has been demonstrated that with the requested changes the dose 
consequences of these events meet the acceptance criteria of 10 CFR 
50.67 and following the provisions of RG 1.183. Thus, the proposed 
amendment will not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois

    Date of amendment request: July 29, 2003.
    Description of amendment request: The proposed amendments would 
allow the licensee to modify technical specifications (TS) to be 
consistent with Technical Specification Task Force (TSTF) Traveler 
TSTF-360, Revision 1, ``DC Electrical Rewrite,'' and to implement new 
actions for inoperable battery chargers, modify certain actions and 
surveillance requirements, relocate certain surveillance requirements 
to a licensee controlled program, and create an administrative program 
for battery monitoring and maintenance to be referenced in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system. The 
proposed changes add actions to specifically address battery charger 
inoperability. This change will rely upon the capability of 
providing the battery charger function by an alternate means (e.g., 
a 125 volts direct current (VDC) portable battery charger or a 250 
VDC portable battery charger) to justify the proposed Completion 
Times. The DC electrical power system, including associated battery 
chargers, is not an initiator to any accident sequence analyzed in 
the Updated Final Safety Analysis Report (UFSAR). Operation in 
accordance with the proposed TS ensures that the DC electrical power 
system is capable of performing its function as described in the 
UFSAR. Therefore, the mitigative functions supported by the DC 
electrical power system will continue to provide the protection 
assumed by the analysis.
    The relocation of preventive maintenance surveillance, and 
certain operating limits and actions, to a newly-created licensee 
controlled Battery Monitoring and Maintenance Program will not 
challenge the ability of the DC electrical power system to perform 
its design function. Appropriate monitoring and maintenance, 
consistent with industry standards, will continue to be performed. 
In addition, the DC electrical power system is within the scope of 
10 CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure the control 
of maintenance activities associated with the DC electrical power 
system. The integrity of fission product barriers, plant 
configuration,

[[Page 59216]]

and operating procedures as described in the UFSAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. This change will rely upon the capability 
of providing the battery charger function by an alternate means 
(e.g., a swing charger or a portable battery charger) to justify the 
proposed Completion Times when a normal battery charger is 
inoperable. The DC electrical power system, including associated 
battery chargers, is not an initiator to any accident sequence 
analyzed in the UFSAR. Rather, the DC electrical power system is 
used to supply equipment used to mitigate an accident.
    The 125 VDC portable battery charger will be utilized as a 
common spare to feed the Division I or Division 2 125 VDC bus of 
Unit 2 or Unit 3. For the 250 VDC system, a full capacity swing 
charger is available for use between the units, and can be aligned 
to any one of the 250 VDC batteries. In addition, the 250 VDC 
portable battery charger can be utilized as a common spare to feed 
the 250 VDC safety related batteries of Unit 2 or Unit 3. This 
portable charger is identical to the existing chargers and is non-
safety related. The output of the portable charger will be capable 
of being connected to any one of the Class IE DC buses for Division 
I or Division 2 of Unit 2 or Unit 3. Allowing the use of a portable 
spare and swing battery chargers will increase the reliability of 
the DC electrical power system. The mitigative functions supported 
by the DC electrical power system will continue to provide the 
protection assumed by the safety analyses described in the UFSAR. 
Therefore, there are no new types of failures that could be created 
by a failure of the portable battery charger. As such, no new or 
different kind of accident or transient is expected by these 
changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new Battery Maintenance 
and Monitoring Program will ensure that the station batteries are 
maintained in a highly reliable manner. The use of a portable 
battery charger will increase the reliability of the DC system 
during periods of normal battery charger inoperability. The 
equipment fed by the DC electrical sources will continue to provide 
adequate power to safety related loads in accordance with analysis 
assumptions. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Senior Counsel, Nuclear; Exelon Generation 
Company LLC; 4300 Winfield Road; Warrenville, IL 60555.
    NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: July 14, 2003.
    Description of amendment request: The proposed change is requested 
to support application of an alternative source term methodology, with 
the exception that Technical Information Document 14844, ``Calculation 
of Distance Factors for Power and Test Reactor Sites,'' will continue 
to be used as the radiation dose basis for equipment qualification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The implementation of alternative source term (AST) assumptions 
has been evaluated in revisions to the analyses of the following 
limiting design basis accidents (DBAs) at Peach Bottom Atomic Power 
Station (PBAPS):

    [sbull] Loss-of-Coolant Accident,
    [sbull] Main Steam Line Break Accident,
    [sbull] Fuel Handling Accident, and
    [sbull] Control Rod Drop Accident.

    Based upon the results of these analyses, it has been 
demonstrated that, with the requested changes, the dose consequences 
of these limiting events are within the regulatory guidance provided 
by the NRC for use with the AST. This guidance is presented in 10 
CFR 50.67 and associated Regulatory Guide 1.183, and Standard Review 
Plan Section 15.0.1. The Alternative Source Term is an input to 
calculations used to evaluate the consequences of an accident, and 
does not by itself affect the plant response, or the actual pathway 
of the radiation released from the fuel. It does however, better 
represent the physical characteristics of the release, so that 
appropriate mitigation techniques may be applied. Therefore, the 
consequences of an accident previously evaluated are not 
significantly increased.
    The equipment affected by the proposed changes is mitigative in 
nature, and relied upon after an accident has been initiated. 
Application of the Alternative Source Term (AST) does not involve 
any physical changes to the plant design. While the operation of 
various systems do change as a result of these proposed changes, 
these systems are not accident initiators. Application of the AST is 
not an initiator of a design basis accident. The proposed changes to 
the Technical Specifications (TS), while they revise certain 
performance requirements, do not involve any physical modifications 
to the plant. As a result, the proposed changes do not affect any of 
the parameters or conditions that could contribute to the initiation 
of any accidents. As such, removal of operability requirements 
during the specified conditions will not significantly increase the 
probability of occurrence for an accident previously analyzed. Since 
design basis accident initiators are not being altered by adoption 
of the Alternative Source Term analyses, the probability of an 
accident previously evaluated is not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed changes). Similarly, it does not 
physically change any structures, systems or components involved in 
the mitigation of any accidents, thus, no new initiators or 
precursors of a new or different kind of accident are created. New 
equipment or personnel failure modes that might initiate a new type 
of accident are not created as a result of the proposed amendment.
    As such the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Safety margins and analytical conservatisms have been evaluated 
and have been found acceptable. The analyzed events have been 
carefully selected and margin has been retained to ensure that the 
analyses adequately bound postulated event scenarios. The dose 
consequences due to design basis accidents comply with the 
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide 
1.183.
    The proposed amendment is associated with the implementation of 
a new licensing basis for PBAPS Design Basis Accidents (DBAs). 
Approval of the change from the

[[Page 59217]]

original source term to a new source term taken from Regulatory 
Guide 1.183 is being requested. The results of the accident 
analyses, revised in support of the proposed license amendment, are 
subject to revised acceptance criteria. The analyses have been 
performed using conservative methodologies, as specified in 
Regulatory Guide 1.183. Safety margins have been evaluated and 
analytical conservatism has been utilized to ensure that the 
analyses adequately bound the postulated limiting event scenario. 
The dose consequences of these DBAs remain within the acceptance 
criteria presented in 10 CFR 50.67, ``Accident Source Term'', and 
Regulatory Guide 1.183.
    The proposed changes continue to ensure that the doses at the 
exclusion area boundary (EAB) and low population zone boundary 
(LPZ), as well as the Control Room, are within corresponding 
regulatory limits.
    Therefore, operation of PBAPS in accordance with the proposed 
changes will not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: Mr. Edward Cullen, Vice President and 
General Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
S23-1, Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 12, 2003.
    Description of amendment request: The proposed amendment is for 
relaxation of the heater acceptance criteria contained in Surveillance 
Requirement (SR) 4.6.6.1d.5, SR 4.7.6.1d.3, and SR 4.7.7d.4 for the 
shield building ventilation, control room ventilation, and controlled 
ventilation area systems, respectively. These SRs are performed to 
verify that heat dissipated by the heaters is within a given band. The 
requested change is to increase the upper limit of the acceptance 
criteria from rated capacity plus 5 percent (%) to rated capacity plus 
10%. No change is proposed for the lower limit of the band of rated 
capacity minus 10%.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The relaxation of the SR acceptance criteria to increase the 
operating band does not alter the way plant equipment is designed or 
operated. The ESF [engineered safety feature] filtration unit 
heating coils will continue to reduce the humidity of the incoming 
air to 70% relative humidity or below. In addition, the air 
temperature will continue to be controlled such that additional 
iodine will not be released into the environment. Thus, the charcoal 
adsorber will continue to meet its design basis and its efficiency 
will not be adversely affected. The effect of the higher heat 
dissipation has also been evaluated and the ignition temperature of 
the charcoal adsorbers is not approached with flow through the 
systems. In addition, the impact of the new acceptance criterion was 
determined not to impact the loading or fuel consumption of the 
emergency diesel generators.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The relaxation of the SR acceptance criteria to increase the 
operating band does not alter the way plant equipment is designed, 
operated, or tested. No possibility for a new or different accident 
or failure mode is introduced by modifying the SR acceptance 
criteria. The proposed change does not affect the functional 
capability of safety-related equipment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The ESF filtration unit heating coils will continue to reduce 
the humidity of the incoming air to 70% relative humidity or below. 
Thus, the efficiency of the charcoal adsorber will not be adversely 
affected. In addition, the impact of the new acceptance criterion 
was determined not to impact the loading or fuel consumption of the 
emergency diesel generators. Therefore, the systems have the same 
capabilities to mitigate accidents as they had prior to the SR 
acceptance criteria change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: August 27, 2003.
    Brief description of amendments: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would 
be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to 
reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated August 27, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed

[[Page 59218]]

by proposed LCO 3.0.4, are no different than the consequences of an 
accident while entering and relying on the required actions while 
starting in a condition of applicability of the TS. Therefore, the 
consequences of an accident previously evaluated are not 
significantly affected by this change. The addition of a requirement 
to assess and manage the risk introduced by this change will further 
minimize possible concerns. Therefore, this change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: L. Raghavan.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: August 25, 2003.
    Description of amendment request: The proposed license amendment 
request would revise Technical Specification (TS) 3.5.1 to incorporate 
TS Task Force 318 for one Low Pressure Coolant Injection (LPCI) pump 
inoperable in each of the two Emergency Core Cooling Systems (ECCS) 
divisions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not affect the LPCI subsystem design or 
function. The change to TS 3.5.1 Condition A with one LPCI pump 
inoperable in both subsystems is more reliable than the current 
configuration allowed by Condition A. The current TS actions require 
entry into shutdown LCO [Limiting Condition for Operation] 3.0.3 for 
this condition. In addition, for an event that does not impact LPCI 
availability the change provides for more injection flow than the 
current TS 3.5.1 Condition A LPCI pump configuration. Review of 
Updated Safety Analysis Report Section XIV-6.0 ``Analysis of Design 
Basis Accidents'' confirms that the LPCI mode of the Residual Heat 
Removal system is not assumed to be the initiator of any previously 
analyzed event.
    Based on the above, NPPD concludes that the proposed TS change 
to TS 3.5.1 Condition A does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical change to the 
plant, add any new equipment or require any existing equipment to be 
operated in a manner different from the present system design.
    Based on the above, NPPD concludes that the proposed TS change 
to TS 3.5.1 Condition A does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed TS change will not reduce the margin of safety. The 
proposed configuration of one LPCI pump in each LPCI subsystem 
represents a more reliable configuration. The current TS actions 
require entry into shutdown LCO 3.0.3 for this condition. In 
addition, for an event that does not impact LPCI availability the 
change provides for more injection flow than the current [LCO] 
requirement which only allows two LPCI pumps in one ECCS subsystem 
to be inoperable for seven days.
    Based on the above, NPPD concludes that the proposed TS change 
to TS 3.5.1 Condition A does not involve a significant reduction in 
the margin of safety.
    From the above discussions, NPPD concludes that the proposed 
amendment involves no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant, 
Van Buren County, Michigan

    Date of amendment request: September 18, 2003.
    Description of amendment request: The proposed amendment would 
revise the limiting condition for operation (LCO) and the associated 
surveillance requirements of Technical Specification 3.4.1, ``[Primary 
Coolant System] PCS Pressure, Temperature, and Flow Departure from 
Nucleate Boiling (DNB) Limits,'' to reflect relocation of the DNB 
limits from the TSs to the Core Operating Limits Report (COLR). These 
DNB limits are for pressurizer pressure, PCS cold leg temperature, and 
PCS total flow rate. The proposed amendment would also revise paragraph 
a of TS 5.6.5, ``Core Operating Limits Report (COLR),'' to reflect the 
addition of ``DNB Limits'' to the COLR. In addition, LCO 3.4.1 would be 
added to items 16 and 17 in TS 5.6.5b, which lists the documents 
approved by the NRC for the analytical methods for which the licensee 
is to use the latest revisions to determine the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 59219]]

    Response: No
    The proposed amendment relocates the primary coolant system 
(PCS) departure from nucleate boiling (DNB) limits to the core 
operating limits report (COLR) and does not involve any change to 
the PCS DNB limits themselves. The proposed amendment does not 
involve operation of any required structures, systems, or components 
(SSCs) in a manner or configuration different from those previously 
recognized or evaluated. The Nuclear Regulatory Commission (NRC) has 
approved all the analytical methods described in Technical 
Specification (TS) section 5.6.5, ``Core Operating Limits Report 
(COLR).'' Relocation of the PCS DNB limits to the COLR will maintain 
existing operating fuel cycle analysis requirements. Any future 
revisions to the safety analyses that require prior NRC approval are 
identified per the 10 CFR 50.59 review process.
    Therefore, the probability of an accident previously evaluated 
will not be increased by the proposed change.
    The consequences of an accident previously evaluated will not be 
increased since the reactor is still protected from violating the 
PCS DNB parameters used in the safety analysis for Palisades Nuclear 
Plant.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment to relocate the PCS DNB limits to the 
COLR would not change or add a system function. The proposed 
amendment does not involve operation of any required SSCs in a 
manner or configuration different from those previously recognized 
or evaluated. No new failure mechanisms will be introduced by the 
proposed change.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment to relocate the PCS DNB limits to the 
COLR will continue to assure that the acceptance criteria 
established in the safety analysis will be met. The safety analyses 
of normal operating conditions and anticipated operational 
occurrences assume initial conditions within the normal steady state 
envelope. The limits placed on DNB related parameters ensure that 
these parameters, when appropriate measurement uncertainties are 
applied, will not be less conservative than those assumed in the 
safety analyses and thereby provide assurance that the minimum 
departure from nucleate boiling ratio (DNBR) will meet the required 
criteria for each of the analyzed transients. The proposed amendment 
does not change the existing PCS DNB limits. Any future revisions to 
the safety analyses that require prior NRC approval are identified 
per the 10 CFR 50.59 review process.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: L. Raghavan.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: September 15, 2003.
    Description of amendment requests: In Technical Specification (TS) 
2.0, ``Safety Limits (SLs),'' Reactor Core SL 2.1.1.2, the proposed 
change would replace the peak linear heat rate SL with a peak fuel 
centerline temperature SL. This change is requested so SL 2.1.1.2 
adequately conforms to 10 CFR 50.36(c)(1)(ii)(A), which requires that 
Limiting Safety System Settings prevent a Safety Limit from being 
exceeded.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not require any physical change to any 
plant systems, structures, or components nor does it require any 
change in systems or plant operations. The proposed change does not 
require any change in safety analysis methods or results. The change 
to establish the PFCT [Peak Fuel Centerline Temperature] as the SL 
is consistent with the Standard Review Plan (SRP) and the SONGS 
Units 2 and 3 licensing basis for ensuring that the fuel design 
limits are met. Operations and analysis will continue to be in 
compliance with NRC regulations.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The SONGS Units 2 and 3 Updated Final Safety Analysis Report 
(UFSAR) Chapter 15 accident analysis for Anticipated Operational 
Occurrences (AOOs) where the peak linear heat rate may exceed the 
existing Safety Limit of 21 KW/ft is the Control Element Assembly 
(CEA) Withdrawal at subcritical and low power startup conditions.
    The accident analyses indicate that the peak linear heat rate 
may exceed the Limiting Safety System Setpoint of 21 KW/ft during 
Control Element Assembly Withdrawal Events at Subcritical and Hot 
Zero Power conditions. The analyses for these AOOs indicate that the 
PFCT is not approached or exceeded. The existing analyses remain 
unchanged and do not affect any accident initiators that would 
create a new accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not require any change in accident 
analysis methods or results. Therefore, by changing the SL from PLHR 
[Peak Linear Heat Rate] to Peak Fuel Centerline Temperature, the 
margin as established in the current license basis remains 
unchanged.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station 
(VCSNS), Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: July 29, 2003.
    Description of amendment request: The proposed change will revise 
Surveillance Requirement 4.0.5 to reflect the deletion of Subsections 
IWP and IWV from Section XI of the 2000 Addenda of American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. This 
change will also result in revising the Technical Specification (TS) 
Bases for 4.0.5, 3/4.4.2 and 3/4.4.6 to reflect the applicability of 
the Code for Operation and Maintenance of Nuclear Power Plants (OM 
Code) to inservice testing activities. TS 4.0.5 is also being revised 
as recommended by NUREG-1492, ``Guidelines for Inservice Testing at 
Nuclear Power Plants,'' April 1995.

[[Page 59220]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to TS 4.0.5 reflects NRC approval of the 
ASME Code [2000 Adenda], in 10CFR50.55a, for the conduct of 
Inservice Testing (IST). The current TS references use of ASME 
Section XI for this testing, which will no longer be applicable for 
the third IST interval. The adoption of an NRC approved test code, 
as required by 10CFR50.55a(f)(4)(ii) will not increase the 
probability of an accident previously evaluated. Testing is 
performed to ensure the operational readiness of pumps and valves to 
perform their safety functions.
    The probability or consequences of accidents previously 
evaluated in the VCSNS FSAR [Final Safety Analysis Report] are 
unaffected by this proposed change because there is no change to any 
equipment response or accident mitigation scenario. There are no 
additional challenges to fission product barrier integrity. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change involves the adoption of an NRC approved 
Inservice Testing Code for the conduct of Operating License mandated 
testing. The adoption of the new Code is required to satisfy 
10CFR50.55a(f)(4)(ii). The new Code enhances plant safety by 
requiring the bi-directional testing of check valves and 
comprehensive pump testing. These changes were incorporated to 
better monitor pumps and check valves for degradation. The adoption 
of the new Code does not create the possibility of a new or 
different kind of accident or malfunction.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of the safety-
related systems and components. A change to the surveillance 
requirement is proposed, but the ASME OM Code is an NRC approved 
standard incorporating inservice testing enhancements not contained 
in ASME Section XI.
    Pursuant to 10 CFR 50.91, the preceding analyses provide a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: September 2, 2003.
    Description of amendment request: The proposed change involves the 
extension from 1 hour to 24 hours of the completion time (CT) for 
Condition B of Technical Specification (TS) 3.5.1, which defines 
requirements for accumulators. Accumulators are part of the emergency 
core cooling system and consist of tanks partially filled with borated 
water and pressurized with nitrogen gas. The contents of the tank are 
discharged to the reactor coolant system if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the 
accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore 
an accumulator to operable status when it has been declared inoperable 
for a reason other than the boron concentration of the water in the 
accumulator not being within the required range. This change was 
proposed by the Westinghouse Owners Group participants in the Technical 
Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is 
supported by NRC-approved topical report WCAP-15049-A, ``Risk-Informed 
Evaluation of an Extension to Accumulator Completion Times,'' submitted 
on May 18, 1999. The NRC staff issued a notice of opportunity for 
comment in the Federal Register on July 15, 2002 (67 FR 46542), on 
possible amendments concerning TSTF-370, including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on March 12, 2003 (68 FR 11880). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated September 2, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The basis for the accumulator limiting condition for operation 
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a 
sufficient volume of borated water will be immediately forced into 
the core through each of the cold legs in the event the RCS pressure 
falls below the pressure of the accumulators, thereby providing the 
initial cooling mechanism during large RCS pipe ruptures. As 
described in Section 9.2 of WCAP-15049-A, the proposed change will 
allow plant operation with an inoperable accumulator for up to 24 
hours, instead of 1 hour, before the plant would be required to 
begin shutting down. The impact of the increase in the accumulator 
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total 
plant core damage frequency (CDF) less than 1.0E-03/yr. The 
incremental conditional core damage probabilities calculated in 
WCAP-15049-A for the accumulator CT increase meet the criterion of 
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using 
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach 
for Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' for all cases except those that are based on 
design basis success criteria. As indicated in WCAP-15049-A, design 
basis accumulator success criteria are not considered necessary to 
mitigate large break loss-of-coolant accident (LOCA) events, and 
were only included in the WCAP-15049-A evaluation as a worst case 
data point. In addition, WCAP-15049-A states that the NRC has 
indicated that an incremental conditional core damage frequency 
(ICCDP) greater than 5E-07 does not necessarily mean the change is 
unacceptable.
    The proposed technical specification change does not involve any 
hardware changes nor does it affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, engineered safety feature (ESF) actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 59221]]

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this 
proposed technical specification CT increase. All safety systems 
still function in the same manner and there is no additional 
reliance on additional systems or procedures. The proposed 
accumulator CT increase has a very small impact on core damage 
frequency. The WCAP-15049-A evaluation demonstrates that the small 
increase in risk due to increasing the CT for an inoperable 
accumulator is within the acceptance criteria provided in RGs 1.174 
and 1.177. No new accidents or transients can be introduced with the 
requested change and the likelihood of an accident or transient is 
not impacted.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be caused as a result 
of the proposed technical specification change. No new failure mode 
has been created and no new equipment performance burdens are 
imposed.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits.
    The basis for the accumulator LCO, as discussed in Bases Section 
3.5.1, is to ensure that a sufficient volume of borated water will 
be immediately forced into the core through each of the cold legs in 
the event the RCS pressure falls below the pressure of the 
accumulators, thereby providing the initial cooling mechanism during 
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable 
accumulator for up to 24 hours, instead of 1 hour, before the plant 
would be required to begin shutting down. The impact of this on 
plant risk was evaluated and found to be very small. That is, 
increasing the time the accumulators will be unavailable to respond 
to a large LOCA event, assuming accumulators are needed to mitigate 
the design basis event, has a very small impact on plant risk. Since 
the frequency of a design basis large LOCA (a large LOCA with loss 
of offsite power) would be significantly lower than the large LOCA 
frequency of the WCAP-15049-A evaluation, the impact of increasing 
the accumulator CT from 1 hour to 24 hours on plant risk due to a 
design basis large LOCA would be significantly less than the plant 
risk increase presented in the WCAP-15049-A evaluation.

    Therefore, this change does not involve a significant reduction in 
a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorneys for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201; Mr. Arthur H. Domby, Troutman Sanders, NationsBank 
Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-
2216.
    NRC Section Chief: John A. Nakoski.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: May 22, 2003.
    Description of amendment request: The proposed amendment revises 
Technical Specification 3.3.2 governing radiation monitoring 
instrumentation to relax restrictions on containment purge valve 
operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The radiation monitors affected by the proposed amendment are 
not potential accident initiators. Adequate measures are available 
to compensate for instrumentation that is out of service. The 
proposed amendment does not affect how the affected instrumentation 
normally functions or its role in the response of an operator to an 
accident or transient. The core damage frequency in the STP [South 
Texas Project] PRA [probabilistic risk assessment] is not impacted 
by the proposed changes. Therefore, STPNOC [South Texas Project 
Nuclear Operating Company] concludes that there is no significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The instrumentation affected by the proposed amendment is not 
credited for the prevention of any accident not evaluated in the 
safety analysis. The proposed amendment involves no changes in the 
way the plant is operated or controlled. It involves no change in 
the design configuration of the plant. No new operating environments 
are created. Therefore, STPNOC concludes the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change has no significant effect on functions that 
are supported by the affected instrumentation. There will be no 
significant effect on the availability and reliability of the 
affected instrumentation. Adequate measures are available to 
compensate for instrumentation that is out of service. Therefore, 
STPNOC concludes the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: August 7, 2003.
    Brief description of amendments: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would 
be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to 
reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated August 7, 2003.

[[Page 59222]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Allen G. Howe.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: July 18, 2003.
    Brief description of amendments: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would 
be eliminated, and Surveillance Requirement 3.0.4 revised to reflect 
the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated July 18, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:



Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of required actions that must 
be performed within the prescribed completion times. The net effect 
of being in a TS condition on the margin of safety is not considered 
significant. The proposed change does not alter the required actions 
or completion times of the

[[Page 59223]]

TS. The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: August 29, 2003.
    Brief description of amendment request: The proposed amendment 
would revise the Updated Final Safety Analysis Report to use the 
reactor building crane for heavy loads up to a total of 117 tons for 
removal and reinstallation activities for the reactor shield blocks 
prior to and during the Units 2 outage D2R18.
    Date of publication of individual notice in Federal Register: 
September 10, 2003.
    Expiration date of Individual notice: October 10, 2003.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: April 2, 2001, as supplemented 
by letters dated January 15, and August 23, 2002, March 28, and August 
19, 2003.
    Brief description of amendment: The amendment identifies the 
conditions under which the inclined fuel transfer system blind flange 
may be removed when primary containment integrity is required (i.e., 
during Modes 1, 2, and 3) and restricts this configuration to no more 
than 40 days per operating cycle. These changes are reflected by (1) 
adding Note 5 for the Actions of Technical Specification (TS) 3.6.1.3, 
``Primary Containment Isolation Valves (PCIVs),'' (2) deleting Note 3 
of TS Surveillance Requirement 3.6.1.3.3, (3) adding a conditional note 
to TS 3.6.1.1, ``Primary Containment--Operating,'' and (4) associated 
TS Bases changes.
    Date of issuance: September 17, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 158.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 13, 2003 (68 FR 
25650). The supplemental letter of August 19, 2003, contained 
clarifying information and did not change the initial no significant 
hazards consideration determination and did not expand the scope of the 
original Federal Register Notice. The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated September 
17, 2003.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: July 31, 2002, and supplemented 
by letters dated March 7 and August 28, 2003.
    Brief description of amendment: The amendment revises Appendix A, 
Technical Specifications (TSs), of the Operating License by adding a 
Surveillance Requirement (SR) to TS 3.2.2, ``Minimum Critical Power 
Ratio (MCPR),'' that requires determination of the MCPR limits 
following completion of control rod scram time testing. The new SR 
provides for the required evaluation necessary to apply faster scram 
times to provide for improved MCPR operating limits.
    Date of issuance: September 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 159.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58637). The supplemental letters

[[Page 59224]]

contained clarifying information and did not change the initial no 
significant hazards consideration determination and did not expand the 
scope of the original Federal Register Notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
September 29, 2003.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: January 14, 2003.
    Brief description of amendment: The amendment revised technical 
specification sections 3.8.9, 3.15.2, 4.12.2, and associated Bases to 
delete the requirements for the reactor building purge air treatment 
system.
    Date of issuance: September 23, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 245.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: (68 FR 10278) March 4, 
2003. The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 23, 2003.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: September 20, 2002.
    Brief description of amendment: The amendment revises the Technical 
Specification (TS) definition of containment integrity to ensure that 
all power-operated valves, relief valves, and check valves are included 
and clarifies the handling of operability and reportability issues 
related to Type III containment isolation valves. The amendment also 
includes minor administrative and editorial changes to improve the 
consistency and clarity of the TSs.
    Date of issuance: September 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 246.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68729). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 2003.
    No significant hazards consideration comments received: No.

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit No. 2, Maricopa County, Arizona

    Date of application for amendment: December 21, 2001, as 
supplemented by letters dated March 13, August 27, August 29, September 
4, September 6, October 11, November 21, December 10, December 23, 
2002, and March 11, June 10, July 25, and August 22, 2003.
    Brief description of amendment: The amendment changes the Unit 2 
Technical Specifications and operating license to support (1) 
replacement of the steam generators and (2) the subsequent operation at 
an increased maximum power level of 3990 MWt, which is a 2.94 percent 
increase from the current 3876 MWt.
    Date of issuance: September 29, 2003.
    Effective date: This license amendment is effective as of the date 
of issuance, and shall be implemented prior to entry into Mode 4 during 
the restart from the Fall 2003 refueling outage.
    Amendment No.: Unit 2-149.
    Facility Operating License No. NPF-51: The amendment revised the 
Technical Specifications and Operating License.
    Date of initial notice in Federal Register: February 19, 2002 (67 
FR 7412). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 29, 2003.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: August 14, 2002, as supplemented 
on March 11, May 16, and May 23, 2003.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) related to reactivity control systems, power 
distribution limits, and special test exceptions.
    Date of issuance: September 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 280.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58640). The supplements dated March 11, May 16, and May 23, 2003, 
provided additional information which clarified the application, did 
not expand the scope of the application as originally noticed, and did 
not change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated September 30, 2003.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: August 7, 2002, as supplemented 
on October 23, 2002.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 6.9.1.8, ``Core Operating Limits Report,'' to update 
the list of documents that describe the analytical methods used to 
determine the core operating limits.
    Date of issuance: September 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
prior to Mode 4 operation of Cycle 16.
    Amendment No.: 281.
    Facility Operating License No. DPR-65: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: September 17, 2002 (67 
FR 58639). The supplement dated October 23, 2002, provided additional 
information which clarified the application, did not expand the scope 
of the application as originally noticed, and did not change the 
staff's original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 25, 2003.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: May 7, 2002, as supplemented on 
January 16, May 27, July 1, and August 21, 2003.
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) 2.2, ``Limiting Safety System Settings'' and 3/
4.3, ``Instrumentation'' to more accurately reflect the existing plant 
design for the Reactor Protection System, the Engineered Safety 
Features Actuation System, and the Radiation Monitoring System 
instrumentation and to provide

[[Page 59225]]

consistency within the associated TS Tables.
    Date of issuance: September 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 282.
    Facility Operating License No. DPR-65: This amendment revises the 
TSs.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42819). The supplements dated January 16, May 27, July 1, and August 
21, 2003, provided additional information which clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 25, 2003.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: August 14, 2002, as supplemented 
December 19, 2002.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) related to Containment Systems. Specifically, the 
revisions: (1) Added clarification to TS 1.7, ``Definitions--
Containment Integrity;'' (2) added clarifying information, as well as 
revised a portion of Surveillance Requirement 4.6.1.1 associated with 
the affected section of TS 3.6.1.1, ``Containment Integrity;'' (3) 
revised TS 3.6.3, ``Containment Isolation Valves,'' that made editorial 
changes, added clarifying information, and added an Action item that 
increased the allowed outage time from 4 hours to 72 hours for 
Containment Isolation Valves in closed systems; and (4) made other 
changes that were clarifying and/or administrative in nature. In 
addition, the TS Bases were revised to address these changes, as 
appropriate.
    Date of issuance: September 29, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days from the date of issuance.
    Amendment No.: 216.
    Facility Operating License No. NPF-49: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61678). The December 19, 2002, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the amendment beyond the scope of 
the initial notice. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated September 29, 2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: December 20, 2001, as 
supplemented by letters dated March 4, 2002, September 12, 2002, 
November 20, 2002, and August 28, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 3.3.2, Engineered Safety Features 
Actuation System Instrumentation.
    Date of issuance: September 10, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 208 and 202.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2002 (67 FR 
12601). The supplements dated March 4, 2002, September 12, 2002, 
November 20, 2002, and August 28, 2003, provided clarifying information 
that did not change the scope of the December 20, 2001, application or 
the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 10, 2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: January 31, 2003, as 
supplemented by letters dated June 12, and September 2, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specifications to incorporate revised means of determining 
the mass of ice in the ice condenser containment.
    Date of issuance: September 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 209 and 203.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003, (68 FR 
18274). The supplements dated June 12, and September 2, 2003, provided 
clarifying information that did not change the scope of the January 31, 
2003, application nor the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 29, 
2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: January 31, 2003, as 
supplemented by letters dated June 12, and September 2, 2003.
    Brief description of amendments: The amendments revise the 
Technical Specifications to incorporate revised means of determining 
the mass of ice in the ice condenser containment.
    Date of issuance: September 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 217 and 199.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003, (68 FR 
18274). The supplements dated June 12, and September 2, 2003, provided 
clarifying information that did not change the scope of the January 31, 
2003, application nor the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 29, 
2003.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: March 20, 2003, supplemented by 
letters dated July 22, and August 5, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications and the licensing basis in the Updated Safety 
Analysis Report to support installation of a passive low-pressure 
injection cross connect inside containment.
    Date of Issuance: September 29, 2003.
    Effective date: As of the date of issuance and shall be implemented

[[Page 59226]]

within 90 days from the date of issuance.
    Amendment Nos.: 335, 335, and 336.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
22745). The supplement dated July 22 and August 5, 2003, provided 
clarifying information that did not change the scope of the March 20, 
2003, application nor the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated September 29, 
2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: August 16, 2002, as supplemented 
June 6, 2003.
    Brief description of amendment: The amendment adds a new Technical 
Specification (TS) requirement to the Pilgrim Nuclear Power Station 
(Pilgrim) TSs consistent with Technical Specification Task Force 
(TSTF)-358. TSTF-358 addresses modifications to requirements for missed 
surveillances consistent with NUREG 1433, Revision 2, ``Standard 
Technical Specification, General Electric Plants, BWR/4'' (STS) 
surveillance requirement 3.0.3. The amendment to the Pilgrim TSs is 
added as TS 4.0.3.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of opportunity for comment in the Federal Register on June 14, 2001 (66 
FR 32400), on possible amendments concerning missed surveillances, 
including a model safety evaluation (SE) and model no significant 
hazards consideration (NSHC) determination, using the consolidated line 
item improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 28, 2001 (66 FR 
49714). The licensee affirmed the applicability of the model NSHC 
determination in its application dated August 16, 2002, as supplemented 
on June 6, 2003.
    In addition, the following statement was added to the TS definition 
of Limiting Condition for Operation (LCO): ``Failure to meet a 
Surveillance, whether such failure is experienced during the 
performance of the Surveillance or between performances of the 
Surveillance, shall be failure to meet the LCO.'' The amendment also 
made administrative changes to add new TS Sections 3.0, ``Limiting 
Condition for Operation (LCO) Applicability,'' and 4.0, ``Surveillance 
Requirement (SR) Applicability,'' into the Pilgrim TSs. New TSs 3.0, 
4.0.1, and 4.0.2 are identified as ``Not Used.'' These changes rectify 
the differences in the format and terminology of the current Pilgrim 
TSs compared to the STS. The associated Bases are also implemented.
    Date of issuance: September 30, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 203.
    Facility Operating License No. DPR-35: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43390). The Commission's related evaluation of the amendment is 
contained in a SE dated September 30, 2003.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: March 11, 2003.
    Brief description of amendment: The amendment revises and relocates 
Surveillance Requirement (SR) 4.0.5 and SR 4.4.9 to the administrative 
section of the Technical Specifications (TS) under sections 6.5.8 and 
6.5.7, respectively. The amendment also relocates TS 3.4.9, ``Reactor 
Coolant System Structural Integrity'' and its Bases to the Technical 
Requirements Manual. Additionally, the amendment extends the Waterford 
3 flywheel volumetric examination interval to ten years.
    Date of issuance: September 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 189.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28851). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 22, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendments: April 19, 2002, and as 
supplemented September 9, 2002, January 3, and July 13, 2003.
    Brief description of amendments: The amendments would revise the 
surveillance frequency of the containment spray system nozzles from 10 
years to ``Following maintenance that could result in nozzle blockage, 
OR Following fluid flow through the nozzles.''
    Date of issuance: September 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 134 and 134.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2002 (68 FR 
40023). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 22, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: October 28, 2002.
    Brief description of amendments: The amendments authorize changes 
to the Updated Final Safety Analysis Report to describe the use of cast 
iron materials in the containment cooling service water and diesel 
generator cooling water systems.
    Date of issuance: September 17, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 201 and 193.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revise the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75875). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 17, 2003.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: December 4, 2001.
    Brief description of amendment: The amendment revises the Davis-
Besse Nuclear Power Station Operating License, Appendix A, Technical 
Specifications (TS) Section 6.9, ``Administrative Controls--Reporting 
Requirements,'' to eliminate the requirement to submit startup test 
reports to the NRC.

[[Page 59227]]

    Date of issuance: September 25, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 258.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43391). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 25, 2003.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: July 18, 2002.
    Brief description of amendments: These amendments revise the 
Technical Specifications regarding the time period that inoperable 
channels of the engineered safety feature actuation system can be in 
the bypassed or tripped condition.
    Date of Issuance: September 30, 2003.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 188 and 132.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2002 (67 FR 
53987). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 30, 2003.
    No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: October 11, 2002, as supplemented by 
letters dated April 21, and July 29, 2003.
    Description of amendment request: The amendment revises the 
Technical Specifications (TS) to eliminate the Power Range Neutron Flux 
High Negative Rate Reactor Trip function from TS 3/4.3.1, ``Reactor 
Trip System Instrumentation,'' TS 2.2.1, ``Reactor Trip System 
Instrumentation Setpoints,'' and their associated Bases. The amendment 
also revises TS 3/4.10.3, ``Physics Tests,'' TS 3/4.10.4, ``Reactor 
Coolant Loops,'' and TS Table 4.3-1, ``Reactor Trip System 
Instrumentation Surveillance Requirements,'' that are associated with 
certain testing activities required during STARTUP operations. The 
revision also rewords the time interval for the Analog Channel 
Operational Test (ACOT) in surveillance requirement (SR) 4.10.3.2. In 
correlation with the revision to extend the ACOT interval in SR 
4.10.3.2, Table 4.3-1 Note 1 is revised. This revision also extends the 
ACOT interval for those Functional Units that reference TS Table 4.3-1 
Note 1. The revision to TS 3/4.10.4 will delete TS 3/4.10.4 in its 
entirety. Additionally, as a result of deleting TS 3/4.10.4, the 
footnote which references TS 3/4.10.4 in TS 3/4.4.1.1 is deleted as 
well.
    Date of issuance: October 1, 2003.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 91.
    Facility Operating License No. NPF-86: Amendment revises the TSs.
    Date of initial notice in Federal Register: November 26, 2002 (67 
FR 70767). The April 21, 2003 and July 16, 2003, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
beyond the scope of the initial notice. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 1, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: September 30, 2002, as 
supplemented July 24 and September 25, 2003.
    Brief description of amendment: The amendment authorizes changes to 
the Updated Safety Analysis Report (USAR) to allow the use of an 
upgraded computer code for design-basis accident containment integrity 
analyses called Generation of Thermal-Hydraulic Information for 
Containment (GOTHIC) version 7.0p2 (GOTHIC 7) with noted conditions.
    Date of issuance: September 29, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 169.
    Facility Operating License No. DPR-43: The amendment authorizes 
changes to the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: October 29, 2002 (67 FR 
66011). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated September 29, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: April 30, 2003.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant, Technical Specification Section 6.3, ``Plant Staff 
Qualifications.'' The amendment updates requirements that have been 
outdated based on licensed operator training programs being accredited 
by the National Academy for Nuclear Training and promulgation of the 
revised 10 CFR 55, ``Operators' Licenses.''
    Date of issuance: October 2, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 170.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34670). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: March 11, 2003, as supplemented 
July 16, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.1.4, ``Rod Group Alignment Limits,'' and TS 3.1.7, 
``Rod Position Indication,'' to add a 1-hour soak time to both TSs to 
allow the control rod drive mechanisms additional time following 
substantial rod motion to reach thermal equilibrium.
    Date of issuance: October 1, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 160 and 151.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18280). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 1, 2003.

[[Page 59228]]

    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: April 15, 2002, as supplemented 
by letters dated September 27, 2002, February 28, 2003, April 25, 2003, 
June 24, 2003, and September 12, 2003.
    Brief description of amendments: The amendments authorize changes 
to the Final Safety Analysis Report (FSAR) Update, together with other 
analyses, design, and procedure changes, to implement the Diablo Canyon 
Power Plant NUREG-0612, ``Control of Heavy Loads at Nuclear Power 
Plants'' program that is required to implement a dry cask Independent 
Spent Fuel Storage Installation (ISFSI).
    Date of issuance: September 26, 2003.
    Effective date: September 26, 2003, and shall be implemented 
following the implementation of the ISFSI. The implementation of the 
amendments include the incorporation into the FSAR Update the changes 
discussed above, as described in the licensee's application dated April 
15, 2002; its supplements dated September 27, 2002, February 28, 2003, 
April 25, 2003, June 24, 2003, and September 12, 2003; and evaluated in 
the staff's safety evaluation attached to the amendment.
    Amendment Nos.: 162 and 163.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
authorized revision of the FSAR Update.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40025) The supplemental letters dated September 27, 2002, February 28, 
2003, April 25, 2003, June 24, 2003, and September 12, 2003, provided 
additional clarifying information, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 26, 2003.
    No significant hazards consideration comments received: No.

PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: April 10, 2003.
    Brief description of amendments: The amendments revise Salem 
Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), Technical 
Specifications (TSs) Table 3.3-1, ``Reactor Trip System 
Instrumentation,'' by modifying the ``Condition and Setpoint'' 
description of permissive interlock ``P-7.'' The phrase ``Turbine 
impulse chamber pressure,'' contained in the ``Condition and Setpoint'' 
description for permissive P-7, is replaced with the phrase ``Turbine 
steam line inlet pressure'' in order to support planned modifications 
to Salem's high pressure turbines.
    Date of issuance: October 1, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 259 and 240.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34672). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 1, 2003.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: September 24, 2002, supplemented 
by letters dated April 8 and May 21, 2003.
    Brief description of amendment: This amendment revises the Action 
Statement and surveillance requirements for the emergency diesel 
generators (EDGs). The proposed changes would revise TS Section 
3.8.1.1, Action b.2 and Action c.2, and TS Section 4.8.1.1, `` AC 
Sources'' and associated Bases Section related to the EDG.
    Date of issuance: September 26, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 164.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002 (67 
FR 68742). The April 8 and May 21, 2003, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the 
application. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 26, 2003.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: December 19, 2002, as 
supplemented by letters dated April 7, May 21, May 30, June 4, 
September 4, and September 12, 2003.
    Brief description of amendments: The amendments revise the licensed 
power level for Hatch, Units 1 and 2 by 1.5 percent from 2763 megawatts 
thermal (MWt) to 2804 MWt. The change is based on the installation of 
the Advanced Measurement Analysis Group, Inc. (AMAG)/Westinghouse 
Crossflow ultrasonic flow measurement instrumentation, resulting in 
improved feedwater flow measurement accuracy. The amendment changes the 
Renewed Facility Operating License (RFOL) and the Technical 
Specifications (TSs) to reflect the increased licensed power level.
    Date of issuance: September 23, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 238 and 180.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revise the RFOL and the TSs.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7821). The supplements dated April 7, May 21, May 30, June 4, 
September 4, and September 12, 2003, provided clarifying information 
that did not change the scope of the December 19, 2002, application nor 
the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 23, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: February 14, 2003, as 
supplemented by letters dated June 5 and August 21, 2003.
    Brief description of amendment: The amendment consists of changes 
to Technical Specification (TS) 5.9.5, ``Core Operating Limits Report 
(COLR).'' The revised TS modifies TS 5.9.5 to add three additional 
methodologies in support of the Westinghouse 17x17

[[Page 59229]]

Robust Fuel Assembly (RFA)-2 fuel design with Intermediate Flow Mixers.
    Date of issuance: September 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
no later than MODE 6 entry following the next refueling outage in the 
fall of 2003.
    Amendment No.: 46.
    Facility Operating License No. NPF-90: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15765). The supplemental letters provided clarifying information that 
did not expand the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 30, 2003.
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: March 6, 2003, as supplemented by 
letters dated July 25, August 29, and September 16, 2003.
    Brief description of amendments: The amendments revise the Final 
Safety Analysis Report (FSAR) and Technical Specification (TS) Bases 
reflecting approval of elimination of response time testing for 
selected Reactor Trip System and Engineered Safety Features Actuation 
System protection channel equipment.
    Date of issuance: September 25, 2003.
    Effective date: As of the date of issuance. The TS Bases shall be 
implemented within 60 days from the date of issuance and the FSAR shall 
be implemented in the next periodic update to the FSAR in accordance 
with 10 CFR 50.71(e).
    Amendment Nos.: 107 and 107.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the FSAR and TS Bases.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18288). The July 25, August 29, and September 16, 2003, supplemental 
letters provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register on 
April 15, 2003 (68 FR 18288). The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated September 25, 
2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 3rd day of October.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-25742 Filed 10-10-03; 8:45 am]
BILLING CODE 7590-01-P